Text

Proposed Tech Spec Section 3/4.3.1, Reactor Trip Sys Instrumentation, Revising Specified Surveillance Intervals & out-of-svc Times,Per WCAP-10271 & Suppl 1 to WCAP-10271
	ML20215H552
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Site:	Wolf Creek
Issue date:	06/16/1987
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ATTACHMENT III PROPOSED TECHNICAL SPECIFICATION CHANGES 8706240143 870616 PDR ADOCK 05000482 p

PDR 1

- Attachment III to ET'87-0223 Page l' of 16

June,16, 1987 3/4.3 INSTRUMENTATION

^

3 /4. 3.:1 REACTOR TRIP SVSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION

' 3. 3.1 As'a minimum, the Reactor Trip Systea instrumentation channels and interlocks of Table 3.3-l'shall be OPERABLE with RESPONSE TIMES as snow Table 3.3-2.

'ADoLICABIL:TY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REOUIREMENTS 4.3.1.1-Each Reactor Trip System instrumentation channel anc interlock and the automatic trip logic shall be demonstrated OPERABLE by tr.e performance of the Reactor Trip System-Instrumentation Surveillance Recuirements specified in Taole 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of eacn Reactor trip function shall be demonstrated to be within its limit at least once per 18. months.

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the

" Total No. of Channels" column of Table 3.3-1.

4 WOLF CREEK - UNIT 1 3/4 3-1

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inttachreent 111 to ET 87-0223 Page 5 of 16 June 16,1987 TABLE 3.3-1 ' Continued)

TABLE NOTATIONS

0nly if the' Reactor Trip System breakers haopen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.

~"The boron dilution flux doubling signal may be blocked during reactor startuo in'accordance with normal operating procedures.

The provisions of Soecification 3.0.4 are not applicable.

1. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint

1. 1. Balow the P-10 L w Set oint Power Rance Neu ron Flux Interlock) Setooint.

1)The applicable MODE and ACTION statements for these c anne s note in

{

Table 3.3-3 are more restrictive and therefere applicable.

l u un s i sT trren i r -==

1 ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPERATION may proteco-orovided the following conditions are satisfied:

a.

The inocerable channel is placed in the tripped condition within houQ b.

The Minrmum Channels OPERABLE requirement is met,

owever, the inoperable channel may be bypassed for up to-e-hours for surveillance testing of other channels per Specification 4.3.1.1: and c.

Either. THERMAL POWER is restricted to less than or equal to 75?, of RATED THERMAL POWER and the Power Range Neutron Flux Trio Setooint is reduced to less than or eoual to 85% of RATED THERMAL POWER within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hoars per Specification 4.2.4.2.

i ACTION 3.With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

l a.

Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 j

Setooint; or j

b.

Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER. restore the inoperabie channel to OPERABLE status crior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

~

ACTION 4 - With the number of OPERABLE channels'one less than the Minimum Channels'0PERABLE recuirement susoend all operations invoiving positive reactivity changes.

U WOLF CREEK - UNIT 1 3/4 3-5 i

1

Attachment III to ET 87-0223 Page 6 of 16 Jun'e 16,1987 1

l TABLE 3.3-1 (Continued) l

{

ACTION STATEMENTS (Continued)

ACTION 5 - a.

With the number of OPERABLE channels one less than the Mini-mum Channels OPERABLE requirement, restore the inoperaole channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reac-tor Trip Breakers, suspend all operations involving positive reactivity changes and verify valves BG-V178 and BG-V601 are i

closed and secured in position within the next hour.

b.

With no channels OPERABLE, open the Reactor Trip Breakers, l

suspend all operations involving positive reactivity changes and verify compliance with the SHUTDOWN MARGIN reouirements f

of Specification 3.1.1.1 or 3.1.1.2, as applicable, within

{

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter, and verify valves BG-V178 and BG-V601 are closed and secured in position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and verified to be closed and secured in position every 14 days.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripoed conoition a.

wit 5f n ho $nd

(

b.

The Minimum hannels OPERABLE requirement is met: however.

the inoperable channel may be bypassed for uo to-0-hours for surveillance testing of other channels per @

Specification 4.3.1.1.

ACTICN 7 Wi tt. U,e i.w.4e-e f OP:RACL chanaci; cnc 1 :: ther the Tot;'

t.;bcr of Channels, STAP,TUI enuer PCUC CPEP"!C" r.-

prc : d-until perfor,;cacc of the next. required ANALCC CP""EL CPO"!I "4 TEST pr;;id;d the inopcrobic channci

pieced 'n Rc t.ip;;;

condition ;;itH r 1 5:ur ACTION 8

- With less than the Minimum Number of Channels OPERABLE. within I

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s)'that the interlock is in its recuired state for the existing plant condition, or apply Specification 3.0.3.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, be in at least HOT STANCSY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for uo to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1 provided the other channel is OPERABLE.

I 1

ACTICH 10 - With the number of OPERABLE channels one less inan tne Minimum Channels OPERABLE recuirement, restore the ineceracie :narnel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or ocen the Reac.0" tric breakers within the next hour.

ACTICN 11 - With the number of CPERASLE char.nels less than t. e Total Num er

~

of Channels, coeration may contin'ue proviced :ne inece a::le channels are placed in the trioped condition i nin 1 Our.

WOLF CREEK - UNIT 1 3/4 3-5

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Attachment III-to ET 87-0223 Page 12 of 16 '

June 16,1987 3

l

.1 TABLE 4.3-1 (Continued)

' \\

TABLE NOTATIONS

0nly if the Reactor Trip System breakers happen to be closed and the control rod drive system is capable of rod withdrawal.

1. Below P-6 (Intermediate Range Neutron Flux-Interlock) Setpoint.

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

j (1)

If not performed in previous ay 1

(2) Comparison of calorimetric to exco. power indication above 15% of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1.

(3) Single point comparison of incere to excore AXIAL FLUX DIFFERENCE above 15%

of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applic-able for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

\\

(5) Detector plateau curves shall be obtained, evaluated and comparea to manu-facturer's data.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provi-sions of Specification 4.0.4 are not applicable for entry into MOCE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS, (S) "ith p:uer gr :ter then er g al to thc intccle d R t;; int th:

m ::

f NALCC C"A=[L CP:RATIC"AL T:07 3 hall censist of.crifying inct

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n

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w.

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.w.

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B hMW 3 4 %W al bWbE y

bi W G E i y a54g lp 6 4 %

f4E 4 I mi4 5 T h We 3IW38%

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" I ms W 'T e

(9)

-~; =.y aurveillance in MODES 3", 4* and 5* shall also incluce verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

"r t " '

surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10 minute pe~riod.

(10) Setpoint verification is not required.

(11) At least once per 18 menths and following maintenance or adjustment of ne Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips.

(12) At least once per 18 months during shutdown, verify that on a simulatec Boron Oilution Doubling test signal the normal CVCS discharge valves -i11 close and the centrifugal charging pumps suction valves from the

.,35T will open within 30 seconds.

(13) CHANNEL CALISP.ATION shall include the RTO bypass loops flow (14) h rate.

Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS (15) The surveillance frequency and/or MODES spectfied for these channels in

(

Table 4.3-2 are more restrictive and, therefore, applicable.

)

WOLF CREEK - UNIT 1 3/4 3-12

Attachment III to ET 87-0223 Page 13 01' 16 June 16, 1987 i

s 3 /.:.3 INSTRUMENTATION EASES E/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEA ACTUAriON SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincioence logic is maintained, (3) sufficient reovndancy is main-tained to permit a channel to be out-of-service for. testing or maintenance, and (4) sufficient system functi'onal capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, anc diversity assumed available in the facility oesign for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.

The Surveillance Requirements specified for these syst' ems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at t

'm m frequencies are sufficient to

(

demonstrate this capability.

NS RT A The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the Distacles are set for each functional unit.

A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within i

the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

the OPERABILITY of a channel when its Trip Setpoint is found to exceed t Allowable Value.

The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.

In Equation 3.3-1, 2 + R + 5 5 TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered.

Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis exclucing those associated with the sensor and rack drift and tne accuracy of their measurement.

TA or Total Allowance is the difference, in percent scan, Detween the Trip Setooint and the value used in the analysis for the actuation.

R or Rack Error is the "as measured" deviation, in percent scan, for the affected channel from tne specified Trip Setpoint.

5 or Sensor Error is eitner 4LF CREEK - UNIT 1 E 3/4 3-1

Attachment III to ET 87-0223 Page-1,4 of 16 Jurle 16,1987 BASES I

REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SY INSTRUMENTATION (Continueo) the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.

Trip Setpoints are the magnitudes of these channel uncertainties. Inherent to the di Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

'i Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.

will happen, an infrequent excessive drift is expected.Being that there is a s in axcess of the allowance that is more than occasional, may be indicative ofRack or se mon serious problems and should warrant further investigation.

The measurement of response time at the specified frequencies provides

-assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumec in the safety analyses.

with response times indicated as not applicable.No credit was taken in the analys Response time may be cemon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response

(

)

time as defined.

Sensor response time verification may be demonstrateo by either:

replacement sensors with certified response times.(1) in place, onsite, or o The Engineered Safety Features Actuation System senses selected plant para-meters and determines whether,or not predetermined limits are being exceeced.

If they are, the signals are combined into logic matrices sensitive to ccmcina-tions indicative of various accidents, events, and transients.

logic combination is completed, the system sends actuation signals to tnoseOnce tne recuirec Engineered Safety Features components whose aggregate function best serves the requirements of the condition.

As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate th'e consequences of a steam line break or loss-of-coolant accident:

Injection pumps start and automatic valves position,.(2) Reactor trip,(1) Safety (3) Feec-water System isolates, (4) the emergency diesel generators start, (5) contain-ment spray pumps start and automatic valves position, (6) contain and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergenc:.

Ventilation System.

w0LF C:.EEK

_ UNIT 1 2 3/4 3-2

~

1 Attachment III to ET 87-0223 Page 15 of 16 June 16,1987 I

(

\\

INSTRUMENTATION BASES Encineered Safetv Features Actuation System Interlocks The Engineered Safety Features Mtuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feecwater valves which were closed by a Safety Injection or High Steam Generator Water level signal, allows Safety Injection block so that components can be reset or tripped.

Reactor not tripped prevents manual block of Safety Injection.

D-11 On increasing pressure P-11 automatically reinstates safety injection actuation on low pressurizer pressure end low steamline oressure anc automatically blocks steamline isolation on negative steamline pressure rate, On decreasing pressure; P-11 allows the manual block of Safety Injection on low pressurizer pressure and low steamline pressure and allows steamline isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI.

/

3/4.3.3 MONITORING INSTRUMENTATION

(

3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated ACTION will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of service for testing or maintenance.

The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions.

Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and acutation of Emergency Exhaust or Control Room Emergency Ventilation Systems.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the The OPERABILITY of this system is demonstrated by irradiating each core.

cetector used and cetermining the acceptability of its voltage curve.

For the purpose of measuring F (Z) or F g

g a M 1 incore n ux map is used.

~

Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in retalibration of the Jxcore Neutron Flux Detection System, ano full incore flux maos or symmetric incore thimbles may be used for monitoring the OUADRANT POWER TILT RATIO when one Power Range Neutronilux channel is inoperable.

WOLF CREEK - UNIT 1 8 3/4 3-3

l l

Attachment III to ET 87-0223 Page 16 of 16 June 16,1987

~

Insert A i

Specified surveillance intervals and surveillance and maintenance outage times have been determined in acordance with WCAP-10271, "Ev aluation of Surveillance Frequencies and Out of Service times for the Reactor Protection Instrumentation System," supplements to that report, and the NRC's Sa fety Evaluation dated Februar y 21, 1985.

Surveillance intervals and out of service times were determined based on maintaining and an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

.

ML20215H552

Text

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